A once-through cycle would require fissile materials as seed (e. g. U-235 or plutonium) each time the reactor is refuelled. To take advantage of the fertile thorium and its capability for generating fissile U-233 in a once-through cycle, extended cycles would be required to permit continued generation of U-233 and its fissioning to sustain reactor operation. In this regard, the excellent refractory properties of thorium oxide make it a good candidate for extended burn-up.

From the perspective of sustainability and resource utilization, however, there is still a dependency on U-235 with a once-through cycle (or plutonium if this is used as fissile material). In fact, if thorium were to be used in today’s thermal reactors, the conversion factor would be less than one but still greater than the conversion factor achieved in a standard uranium-plutonium cycle. For example, in a LWR using thorium-based fuel, the conversion factor is 0.7 (compared to 0.6 for uranium-plutonium fuels) and may reach easily 0.8 or even 0.9 in other types of reactors such as PHWRs or HTRs. A self-sufficient equilibrium thorium cycle, i. e. a conversion factor equal to or greater than 1, can even be reached in some thermal reactors. Examples are the Shippingport LWR reactor (see Section 8.2.1), CANDU-type reactors and, especially, molten salt reactors (MSR). In MSRs, U-233 breeding is promoted by keeping burn-up and specific power low (which entails an economical penalty) and by continuous removal of Pa-233 from the core by on-line reprocessing. Breeding in a thermal spectrum is not possible with the uranium-plutonium cycle so that this represents a real advantage of the thorium cycle. This is important because of the more favourable characteristics of thermal reactors compared to fast neutron reactors, e. g. lower fissile inventory in the core and, probably, a lower investment cost.

Various studies have investigated the use of thorium in thermal reactors, since many combinations of fuel cycles are possible with a mix of various types of reactor, operating as symbiotic systems. It transpires that thorium can be mixed with four types of fissile material:

• Highly enriched uranium (> 90% of U-235), called the ‘Th-HEU’ cycle. This was the reference fuel for HTR reactors in the 1970s (in the USA and in Germany). But today use ofthis fuel would raise serious proliferation concerns.

• Mid-enriched uranium (20%), which is called the ‘Th-MEU’ or sometimes ‘denatured’ cycle. The underlying idea is to disable direct use of uranium material for nuclear weapons fabrication.

• Plutonium, whatever its origin and isotopic composition, called the ‘Th-Pu’ cycle.

• Uranium-233, when available in large amounts after reprocessing of thorium — based spent fuel.

To summarize, the results of these studies show that thorium use in non-breeder thermal reactors would allow a global saving in uranium usage from a few tens of per cent to a maximum of roughly 80%, when equilibrium of the reactor fleet is reached. The precise figure depends on reactor types (and reactor type combinations) and recycling options.

With regard to the use of thorium in fast neutron reactors (FNRs), a number of studies (performed particularly in Russia for the BN-800 reactor, but also in France and elsewhere in Europe) demonstrated the possibility of achieving self­sufficiency in a Th-232/U-233 fuel cycle, that is to say achieving a conversion factor greater than one. However, Th-232/U-233 fuel performance regarding breeding in FNRs is not as good as uranium-plutonium fuel performance. For example, thorium-based FNRs need very large material inventories in the blankets to achieve negative feedback reactivity effects and a conversion factor greater than 1. The main reason is that plutonium has an eta factor (see Section 8.1.2) slightly better (1.33) than that of U-233 (1.27) for fission by fast (as opposed to thermal) neutrons. Another reason is that the fission cross section of thorium in the fast range is much lower than that of U-238 (one third or so). In summary, the use of a thorium cycle in FNRs is not very attractive, though there are claims that, for sodium-cooled fast neutron reactors, this leads to a reduced positive sodium void coefficient when compared to the standard uranium-plutonium core.5

Globally, if thorium were to be intensively used in non-breeder thermal reactors in closed cycle (i. e. U-233 recycling) the world’s fissile resources would be increased by around a factor 2 or maybe more in the very long term (provided that enough natural uranium is available to sustain such a cycle). If breeder fast reactors were intensively used with uranium-plutonium fuel (and with a conversion factor at least equal to 1), the energy potential of uranium natural resources would be multiplied by a factor 50 to 100. In that case, thorium breeding would multiply again this already huge energy potential by an additional factor 2 or so (depending on available thorium resources).

The core of a power reactor must be designed to safely, reliably and economically produce heat and transfer it to produce steam. In addition to controlling the processes, it must be capable of being refuelled and maintained. The overall process is complex involving neutronics, thermal hydraulics and structural engineering while at all times satisfying the three fundamental safety principles discussed in Section 10.9. The application to reactor core design is discussed in an IAEA guide (IAEA, 2005).

The design of the core is also key to the economics of the plant. Efficient fuel utilisation helps reduce fuel costs while the ability to rapidly refuel the plant improves availability. Both PWRs and BWRs have to be shutdown to be refuelled. To facilitate this, the reactor vessel is set into a refuelling canal. This consists of a structure within the containment, which is steel lined and can be flooded once the RPV head is removed. This then provides both shielding and cooling and is connected to the fuel pool to allow the transfer of fuel assemblies to the fuel pond.

In some designs the fuel pond is in the containment building and is directly connected to the refuelling canal. In others it is in a separate building and so fuel is transferred through a tube between the buildings, remaining underwater during the whole process. The use of fuel ponds inside the containment is conducive to fast refuelling but requires a containment designed to be accessible during normal operation.

Magnox stations generate a number of waste streams characteristic of their type.

‘Mobile’ wastes comprise mainly resins used for pond and waste water treatment; sludges from backwashing of filters, especially those used in the pond water clean-up plant and composed largely of hydrides of magnesium and aluminium, and tritiated desiccants from the coolant treatment plant.

Solid wastes include activated components (e. g. flux-flattening bars, fuel grabs, control rod assemblies); fuel element debris (FED) from the removal of fuel splitters or lugs; and in some reactors graphite sleeves or struts forming part of each fuel element. At some sites, the solid waste vaults contain a very heterogeneous mix of components giving major challenges to processing and packaging.

On many sites, the pond may have become heavily contaminated with fuel corrosion products due to extended storage of fuel or poor control of pond water chemistry. This can result in fission product/fissile contamination of components in contact with the pond water, including skips, FED and filtration plant.

Waste disposal of intermediate level wastes (ILW) in the UK is mostly through encapsulation in cementitious grout in stainless steel packages, for eventual despatch to an ILW repository. The lack of a currently available ILW repository has led to the construction of on-site stores for ILW packages. Low level wastes (LLW) are sent from site to the UK’s Low Level Waste Repository although other options are being explored, including controlled landfill for lower activity LLW, recycling and smelting of contaminated steels and on-site disposal by burial of LLW.

At Dungeness A, FED, which comprises mostly Magnox material, has been dissolved in carbonic acid and routed to a liquid effluent stream that, after a clean-up, is discharged to the sea. The Magnox itself is a low specific activity material, although higher activity components (such as Nimonic springs) and particulates (e. g. pond sludge containing fuel particles) may be associated with it and require separation. Similar dissolution plants are being considered at other sites.

In the UK, the current philosophy for decommissioning Magnox reactors is to defer final demolition:

• remove the fuel

• decommission plant outside the bioshield, and empty waste vaults

• put the reactor into a period of care and maintenance

• demolish the remaining structures

The advantage of the long period of care and maintenance is that shorter-lived radionuclides (such as the strong gamma emitter Co-60, half-life 5.27 years) will have largely decayed, until external dose rates are dominated by much longer — lived nuclides such as Nb-94 and Ag-108. After 100 years the activity level of Co-60 will have dropped by a factor of a million, enabling manned working in the reactor core. There is little benefit to radiation protection from further delay to final decommissioning.

In France, more rapid programmes for Magnox decommissioning are being considered, using remote dismantling in air or in a water-flooded reactor.

Ceramic fuel pellets consist of fuel grains with a range of grain sizes. The initial grain size distribution is determined by the manufacturing process and sintering conditions. If the fuel temperatures are high enough during irradiation — greater than ~1200 °C in UO2 fuel (Ainscough et at., 1973) — the smaller grains shrink while the larger grains grow in a process known as equiaxed grain growth (since there is no preferential direction for the growth). The net result is an increase in mean grain size. The driving force for the equiaxed grain growth process is provided by the stresses induced at the grain boundaries due to their radius of curvature (Olander, 1976). At still higher temperatures — greater than ~1800 °C in oxide fuel (Olander, 1976), which is generally only achieved for CANDU and fast reactor fuel — the fuel pores become mobile and migrate up the radial temperature gradient via an evaporation-condensation mechanism. In doing so, they create grains elongated in the radial direction — the process is thus known as columnar grain growth. The pores that reach the pellet centre form a central void in solid pellets, or enlarge the bore of annular pellets. The migration of pores causes additional densification to that described above. It also causes plutonium redistribution in fast reactor oxide fuel (Bailly et at., 1999). The increase in fuel thermal conductivity due to removal of porosity, together with the central void formation and pellet bore enlargement, cause a significant reduction in fuel centreline temperature — of the order of several hundred degrees centigrade in fast reactor fuel (Olander, 1976).

The HABOG nuclear waste interim storage plant in the Netherlands is a unique concept. It is managed by the COVRA (The Central Organization for Radioactive Waste).14 In the design of the facility special attention was paid to the long storage time of at least 100 years. It stores both nuclear spent fuel and other types of HLW. The facility was designed and built to high safety standards that include resistance to fire, earthquakes, explosions and direct aircraft crashes. This is the only storage facility that has received a licence for 100 years of operation. In addition to a focus on safety, the designers also paid attention to the artistic merits of the facility and its visual appearance, which contributes to its public acceptability. The building’s exterior will be painted every 20 years changing from the original orange to lighter and lighter shades until it becomes white to symbolize the decrease in stored activity. The aim of this approach is to reduce the activity of

the waste through decay over a 100 year period, prior to disposal. The facility will contain canisters with vitrified HLW from reprocessing, canisters with fuel elements from research reactors and drums with high-level waste. This facility is an example visionary approach to spent fuel and radioactive waste management, which could hopefully gain wider application.

Technically, fast neutron reactors provide an elegant solution to the recycling question because all plutonium isotopes can be fissioned by fast neutrons. This allows the isotopic quality of the plutonium to be maintained or improved. Most fast reactors use plutonium as their driver fuel, with easily enough neutrons being produced to sustain the chain reaction. Fission of plutonium-239, for example, produces 25% more neutrons than uranium-235. This means that there are enough neutrons (after losses) not only to maintain the chain reaction but also to convert U-238 into more Pu-239 continuously. Such breeding is also possible in thermal reactors, of course, but not so easily or effectively.

Fast reactor technology is important in long-term considerations of world energy sustainability and they have also been suggested as vehicles for burning ex-military plutonium, about which there is international concern. In economic terms, however, much depends on the value of the plutonium fuel, which is bred and used and this, in turn, relates to the cost of fresh uranium.

Countries in the second group will have, in addition to disused sealed sources and NORM, three main types of radioactive waste: (a) resulting from the operation of nuclear power plants; (b) produced during decommissioning; and (c) spent fuel or, in those cases where fuel is sent away for reprocessing, vitrified fission products.

The first type, NPP operational waste, typically consists of low-level waste (LLW) and short-lived intermediate-level waste (ILW-SL). LLW mostly consists of contaminated items such as discarded items of plant, clothing and paper. Short­lived ILW mostly arises from operations to clean up and recycle process water. These produce spent ion-exchange resins and (especially in former Soviet countries) evaporator concentrate (also known as evaporator bottoms or salt cake). LLW and ILW-SL are often grouped together under a single heading designated by the acronym LILW.

Radioactive decommissioning wastes fall into two categories: contaminated items and activated metals. The former typically consists of cooling circuit components, such as steam generators and pipework that have not been exposed to a neutron flux. Where the geometry is sufficiently simple, it may be possible to decontaminate these components so that the materials may be released for recycling. Activated metals mostly occur in the reactor pressure vessel and its contents where neutrons have been captured by stable isotopes of metals such as iron, nickel and cobalt to produce radioactive species. In terms of activity, the activated wastes constitute by far the greater part of the total.

Depending on national policy, spent fuel may be considered to be either a waste or a resource. In the first case, the fuel elements themselves will be disposed; this is known as ‘direct disposal’ and is followed, for example, in Sweden, Finland and the USA. If spent nuclear fuel is thought of as a resource (as it has been in Belgium and the Netherlands) then it will probably be sent abroad for reprocessing. The recovered uranium and plutonium may be returned as mixed oxide (MOX) fuel while the waste products are packaged as vitrified fission products. Both spent nuclear fuel and vitrified wastes generate significant quantities of heat. One of the advantages of reprocessing spent fuel is the considerable reduction in the volume of waste, albeit with a corresponding increase in specific activity and hence, heat output. Because most of the platonium has actinides have been removed, vitrified waste decays more rapidly — in terms of both heat output and radiotoxicity — than spent fuel and this is a significant benefit to the long term safety case.

Three types of exposure situations are intended to cover the entire range of exposure situations. The three situations are:

1 Planned exposure situations, which are situations involving the planned introduction and operation of sources of radiation. (This type of exposure situation includes situations that were previously categorized as ‘practices’.)

2 Emergency exposure situations, which are unexpected situations such as those that may occur during the operation of a planned situation, or from a malicious act, requiring urgent attention.

3 Existing exposure situations, which are exposure situations that already exist when a decision on control has to be taken, such as those caused by natural background radiation.

Nuclear safety and security are very much the responsibility of owners or operators of a nuclear facility in close cooperation with national authorities.[6] This is laid down in the first principle of the Fundamental Safety Principles of the Safety Standards to protect people and the environment (IAEA, 2006/3).

The International Nuclear Safety Group (INSAG) has considerable experience in regulatory organizations, research, academic institutions and the nuclear industry and provides advice and guidance on nuclear safety approaches, policies and principles (IAEA, 2011/3). In particular, INSAG will provide recommendations and opinions on current and emerging nuclear safety issues to the IAEA, the nuclear community and the public. INSAG has issued a large number of documents, particularly one in 2010, on the interface between safety and security at nuclear power plants (IAEA, 2010/4). The group requests that states ‘set up an appropriate legislative and regulatory framework to ensure control of nuclear power plants, as well as of the transport and uses of nuclear material that present a radiological risk and thus require safety and security provisions’. The group continues to require the necessary competence and authority for the regulator ‘in both the safety and security fields’ and to provide the necessary ‘authority, competence and the financial and human resources necessary to accomplish their tasks’.

I. CROSSLAND, CrosslandConsulting, UK

Abstract: This chapter traces the rise, fall and possible resurgence of nuclear power through the years from the discovery of atomic fission to the present day. It describes how the technology was discovered and developed — first for the purpose of waging war and then for commercial electricity generation. It explains how concerns over nuclear proliferation and safety produced a period of diminishing public confidence in which, paradoxically, there was increased reliance on the technology for electricity production. Finally, it describes how fears over man-made climate change caused many states to turn towards nuclear power only for some to execute a U-turn after the Fukushima accident. This is described by reference to the examples of France, Sweden, California and Germany, all of whom aim to meet the challenge of large scale reductions in greenhouse gas emissions, albeit through different strategies.

Key words: separation and purification of uranium isotopes, nuclear weapons, nuclear power, public opinion on nuclear energy, nuclear fuel reprocessing.